Stress corrosion cracking occurs in a material due to a combination of a corrosive environment and tensile forces placed on the material. Cracking can be induced in materials in different ways including cold forming, welding, grinding, machining, and heat treatment as well as other physical stresses placed on the material. Stress corrosion cracking in nuclear reactor environments is a significant phenomenon that must be carefully monitored for successful operation of a nuclear power plant facility. Without careful monitoring for PWSCC, material defects may begin and may ultimately damage the material. If cracking continues, the materials may be damaged to such an extent that the materials must be removed from service and replaced. In the nuclear reactor environment, such replacement of components is extremely undesirable due to radiological concerns related to worker and facility safety, as well as overall plant economic concerns.
In Boiling Water Reactors (BWRs), different methodologies are used to limit corrosion on reactor water systems. Some methodologies include application of hydrogen water chemistry to limit the overall nuclear reactor environment of these water systems to a more reducing state. The application of hydrogen water chemistry has significant drawbacks, however, in that radiation levels in systems connected to the reactor often increase dramatically, posing a significant risk for workers and equipment. Other methodologies relate to placement of noble metals on the reactor water systems in order to limit the amount of voltage difference between differing reactor water system areas as defined in the standard hydrogen electrode scale. Through experimentation it has been found that values above approximately −0.230 to −0.300 V result in stress corrosion cracking. The placement of noble metals such as iridium, platinum, palladium and rhodium in key corrosion-prone positions has been found to help in limiting damage to reactor water systems by decreasing these harmful values. In new construction, the components may be coated with these metals, thereby providing protection. For nuclear power plants already in operation this alternative is not practical as the components in question must be removed from service and replaced with new components.
In an effort to increase the corrosion resistance of existing nuclear power plant facilities, injection of noble metals into the reactor coolant water stream itself has been found to help improve resistance to stress corrosion cracking. The noble metals are passed into the reactor coolant water stream where these metals coat the insides of various components in the nuclear power plant environment. The noble metals further deposit on an outside layer of the system to be protected. The noble metals decrease the electrochemical corrosion potential of the systems and therefore help to protect the overall system integrity.
Zinc, in the form of zinc oxide, zinc acetate, or zinc borate, has been used in the commercial nuclear industry for a number of years in order to reduce radiation dose rates of system components in both PWRs and BWRs. In order to achieve these dose reduction benefits, zinc has been applied at concentrations of between 5 to 10 ppb in BWRs and 3 to 8 ppb in PWRs.
Laboratory testing has demonstrated that applying high concentrations of zinc (≧10 ppb) may be beneficial in reducing the initiation of PWSCC in PWRs. Some operating United States PWRs currently use zinc at concentrations of 20 to 40 ppb in the reactor coolant, which is believed by the industry, including the Electric Power Research Institute (EPRI), to achieve PWSCC protection; however, no definitive evidence of the magnitude of benefit, if any, has been available to the industry.
Analysis of laboratory test data in open literature indicates that certain types of addition of zinc to the primary water systems in nuclear facilities may be expected to provide a reduction in the initiation rate of PWSCC in Alloy 600 components. A combined project conducted by Airey et al. in 1996 and Angell et al. in 1999 was performed in two test phases. The Phase 1 test consisted of exposing reverse U-bends (RUBs), bent beams (6% and 12% plastic strain) and pre-cracked compact wedge open loading (WOL) samples from various heats of Alloy 600 and Alloy 690 materials. During Phase 1, the specimens were exposed to simulated PWR conditions for a 12-month fuel cycle (1,200 ppm B, 2.2 ppm Li, and 25-50 cc/kg H2). The specimens were exposed in two autoclaves for up to 7,500 hours at 350° C. (662° F.). Zinc was injected into one of the autoclaves, with a target concentration of approximately 40 ppb in the effluent. In order to achieve this target concentration, an initial dose of 100 ppm was used to condition the system.
After 7,500 hours of exposure in the autoclave the RUB specimens exhibited a low incidence of cracking wherein only two zinc specimens and one control specimen cracked. This number of cracked specimens was below an anticipated number of cracked specimens, however the test results did not allow the researchers to draw any meaningful conclusions. An additional result of the testing was that the bent beam specimens tended to exhibit surface crazing, rather than defined cracking. The crazing was attributed to a cold worked surface layer present on the original material. The testing results of the WOL specimens indicated that zinc addition had no effect on crack growth rate.
The above experimental results indicated that zinc injection had little impact on cracking of metal specimens undergoing test conditions. The researchers also observed that the oxide film thickness on specimens exposed to zinc were thinner, 35 nm versus 230 nm on the control specimens. This led the researchers to question whether the stress was too high on the WOL specimens (loaded to 40 MPa√m) to identify an effect of adding zinc. Researchers, however, planned a second set of tests, Phase 2.
Under the Phase 2 testing, the same types of specimens and material heats were used. WOL specimens loaded to 25 MPa√m and 40 MPa√m were included in the test matrix. Additionally, the chemistry environment was modified to reflect an 18-month fuel cycle (1,800 ppm B, 3.5 ppm Li, and 25-50 cc/kg H2).
The result of the Phase 2 tests showed a benefit in the reduction of crack initiation for the Alloy 600 materials from the addition of zinc at high dose concentrations. The RUB samples (results provided in Table 1) showed fewer specimens cracked (up to 67% fewer) in the autoclave containing 40 ppb zinc. Additionally, it was noted that none of the 16 bent beam specimens exposed to zinc cracked, while 6 of the 16 control specimens cracked. These results are provided in Table 2. These results led the researchers to conclude that high levels of zinc addition had a definite impact on PWSCC initiation. Researchers did not, however, explore the effects of low levels of zinc addition.
TABLE 1RUB Test Conditions and Results (662° F.) by Angell et al.Phase 1, simulating beginning of cyclePhase 2, simulating beginning ofwater for a 12-month fuel cyclecycle water for an 18-month fuel cycleWithout zincWith zincWithout zincWith zincZinc—40 ppb—40 ppbTemperature350° C. (662° F.)350° C. (662° F.)Hydrogen25-50 cc H2/kg H2O or 0.1 MPa25-50 cc H2/kg H2O or 0.1 MPaB1200 mg/kg B as H3BO31800 mg/kg B as H3BO3Li2.2 mg/kg Li as LiOH3.5 mg/kg Li as LIOHpH6.75 at 292° C. and 7.10 at 350° C.6.75 at 292° C. and 7.10 at 350° C.Phase 1Phase 2Without zincWith zincWithout zincWith zincSpecimen I.D.Heat7500 hours7500 hours5500 hours7500 hours5500 hours7500 hoursAlloy 60096834, (c)1/62/63/6 (a)3/6 (a)1/6 (a)1/6 (a)StudsvikAlloy600MAAlloy 690752245, (c)0/60/60/4 (a)0/4 (a)0/4 (a)0/4 (a)StudsvikAlloy690TTAlloy 600Not——1/4 3/4 0/4 2/4 Westing houselisted(a) Include two uncracked specimens carried forward from Phase 1. However, for Alloy 600, it was not mentioned if the cracked ones in Phase 2 included the ones from Phase 1. For Alloy 690, the maximum specimen exposure time without failure was 15,000 hours.(b) x/y—x is the accumulated number of specimens cracked; y is the total number of RUB specimens tested.(c) The descriptions on the RUB specimen I.D. were vague and could not be associated directly with the chemical composition of the Alloy 600 and Alloy 690 SG tubes listed. Hence, the specimen heat number was an educated guess.
TABLE 2Alloy 600 Phase 2 Bent Beam ResultsSPECIMENSPLASTICCRACKED/EXPOSEDZINCSTRAINSUR-27505500ADDITIONHEAT(%)FACE*HOURSHOURSYESA6AM0/40/4YESB6AM0/40/4YESA12EP0/40/4YESB12EP0/40/4NOA6AM0/41/4NOB6AM0/40/4NOA12EP1/44/4NOB12EP0/41/4*AM—as machined; EP—electropolished
The results of the WOL crack propagation specimens, however, were consistent with Phase 1 in that no benefit from zinc injection was found, as shown in FIG. 1. Analysis of the fracture surfaces of the control and zinc-exposed specimens revealed no discernable differences in the oxides. These results led the researchers to conclude that zinc most likely was not being transported to the crack tip and therefore had no impact on crack propagation.
Tests were also performed by Kawamura et al. in 1998 and 2000 using mill annealed Alloy 600 tubing and plate in order to evaluate zinc effects on both PWSCC initiation and propagation. Initiation was studied using slow strain rate tests (SSRT) of tubing material at 360° C. (680° F.) in water containing 50 ppm B, 2.2 ppm Li, 25 cc H2/kg H2O, and (for zinc tests) 10 ppb Zn. Propagation was studied using double cantilevered beam (DCB) specimens wedge loaded from <10 to >70 MPa√m.
In many cases, the test specimens were pre-filmed by exposing them to water containing 50 ppm B, 2.2 ppm Li, and 25 cc/kg H2 (both with and without 10 ppb Zn) for up to 2,000 hours prior to testing. Later testing revealed little difference between specimens that were pre-filmed and those that were not.
The SSRT results were reported in terms of “fracture ratios” (i.e., fraction of specimen that showed PWSCC failure). The testing showed a sharp decrease in fracture ratios as a function of zinc concentration between 0 and 10 ppb zinc (see FIG. 2) and a decrease by a factor of two for specimens exposed to zinc; 10-15% versus 20-30% in the control specimens (see FIG. 3).
The crack propagation tests were monitored by periodically removing and fracturing some of the specimens and measuring crack advancement. The results of these tests indicated that crack rates (most likely also including the initiation times into the calculation) were approximately 10 times lower for the zinc environments as provided in FIG. 4. It should be noted that in all cases, the crack growth rates were low. It should also be noted that, contrary to previous studies the researchers found chromite spinel oxides, which can incorporate zinc, in the crack tip.
Research reports have been provided to the Electric Power Research Institute by the Nuclear Power Engineering Corporation (NUPEC) in which the effect of zinc additions on PWSCC have been evaluated. Some of the project details have been published in the open literature. A materials integrity test was performed in a large loop specifically designed and constructed in generating these reports. The types of testing included SSRT, constant load RUB tests, and constant strain RUB tests.
Slow Strain Rate Tests
Slow strain rate testing was performed at 370° C. (698° F.) with a strain rate of 0.5 μm/minute. Three SSRT environmental conditions were used, as shown in Table 3.
TABLE 3Slow Strain Rate Test ConditionsTEST NO.BORON (PPM)LITHIUM (PPM)ZINC (PPB)12802.010218003.510318003.50
Four specimens were included in each SSRT environment. All specimens were prefilmed in primary water conditions containing 10 ppm zinc. The reported fracture ratios were all approximately 10% as shown in FIG. 5, indicating that the presence of 10 ppb zinc had no effect on the PWSCC susceptibility in these B and Li environments.
Constant load testing of Alloy 600 MA and Alloy 600 TT tubing was performed at 340° C. (644° F.). Alloy 690 TT tubing was also included and tested at 360° C. (680° F.). All specimens were strain hardened prior to testing at 60 kg/mm2 and the applied testing load was equivalent to a tensile stress of 588 MPa. The Alloy 690 TT material was not prefilmed; the Alloy 600 TT material was prefilmed in primary water with the addition of 10 ppb zinc; and the Alloy 600 MA material was prepared in three conditions: a) without prefilming, b) prefilmed in primary water only, and c) prefilmed in primary water with the addition of 10 ppb zinc. The prefilming was performed after the strain hardening treatment. The Alloy 600 MA material results are the only ones reported in the literature and are provided in Table 4. The use of prefilming made little to no difference in the test results. Also, the testing performed in the 10 ppb zinc environment (Environment B) appears to show only marginal improvement in the failure times.
Constant strain tests were performed using RUB specimens made from both Alloy 600 MA and Alloy 600 TT materials. Alloy 690, Alloy X-750, Type 316, and Type 304 materials were also included, but the data were not presented in the references. A matrix of environmental test conditions, strain levels and prefilming was employed as shown in Table 5.
The test results indicate that for Alloy 600 MA materials, prefilmed specimens tend to crack more than non-prefilmed specimens, particularly under 5% strain. Higher strain conditions showed no beneficial effect of zinc addition. The test results for Alloy 600 TT material appeared to show a small improvement in PWSCC resistance in the zinc environments. The authors concluded from these test results that PWSCC susceptibility of Alloy 600 MA and TT materials was essentially the same or somewhat lower in a 10 ppb zinc environment compared to a water environment of a typical primary water system in a nuclear power plant.
TABLE 4Constant Load Test Results for Alloy 600 MA MaterialENVIRONMENT A*ENVIRONMENT B**PREFILMSPECIMENRUPTURESPECIMENRUPTURECONDITIONNO.TIME, HRNO.TIME, HRNO ZINC6M-30189876M-31183896M-30288626M-31257256M-30384396M-313>92286M-30486896M-31473676M-305>92286M-315>92286M-30657986M-316>92286M-30778856M-317>92286M-308>92286M-31891296M-309>92286M-319>92286M-31050966M-320921410 PPB6M-32192026M-331>9228ZINC6M-32289176M-33271916M-32369856M-33368366M-32449386M-334>92286M-32543376M-335>92286M-32667046M-33647196M-32764196M-33747896M-32880766M-33889346M-32976666M-33982066M-33048506M-3407650*Environment A: 280 ppm boron and 2.0 ppm lithium**Environment B: 280 ppm boron, 2.0 ppm lithium, and 10 ppb zinc
TABLE 5Constant Strain RUB Testing of Alloy 600 MaterialsENVIRONMENT A+ENVIRONMENT B+ENVIRONMENT C+PREFILMPRE-STRAIN LEVELPRE-STRAIN LEVELPRE-STRAIN LEVELMATERIALCONDITION*5%10%15%5%10%15%5%10%15%ALLOYNONE1/1010/1010/101/1010/109/103/108/10 8/10600 MAPW4/1010/10ALL #4/1010/10ALL #8/1010/10 10/10PW + 107/1010/10ALL #5/1010/10ALL #———PPB ZNPW + 50——————9/109/1010/10PPB ZNALLOYPW—————————600 TTPW + 100/10 0/10 5/100/10 0/103/10———PPBZNPW + 50——————0/101/10 4/10PPB ZNNote:Results for 0 and 10 ppb zinc are after 9228 hours; results for 50 ppb zinc are after 5005 hours; and “All #” indicates all specimens cracked at 5010 hours*Prefilmed 2000 hours in primary water (PW) environment with or without zinc+Environment A: 280 ppm boron, 2.0 ppm lithium at 320° C.; Environment B: A + 10 ppb zinc; and Environment C: A + 50 ppb zinc
Zinc addition has been applied to a number of PWRs in the United States and abroad for the purposes of radiation source term reduction (5 ppb Zn in the reactor water) and PWSCC initiation reduction (20 to 40 ppb Zn in the reactor water). Several studies of these programs, including the results of the field applications, can be found in various EPRI topical reports. None of these reports, however, identified any relationship between the addition of zinc and reduction in pressurized water reactor stress corrosion cracking.
Limitations of Background References
The laboratory data recited above on the effect of low-concentration zinc addition (≦10 ppb in the reactor coolant) regarding the initiation of PWSCC is not comprehensive. Because of the large costs and schedule requirements of zinc addition studies, the various testing programs contain fragmentary information about zinc and its effects on water systems of a nuclear reactor water system.
The field application studies of zinc addition, furthermore, have focused on two “known” applications of zinc. First, zinc has been applied at concentrations of approximately 5 ppb in the reactor coolant, and the resulting effects on the plant radiation fields (generally secondary systems) have been studied. Secondly, zinc has been applied at a concentration of 20 to 40 ppb, and the effect on the initiation rate of PWSCC has been studied, although a definitive quantification of the measure of improvement has not been determined.
There are two major shortcomings of the current knowledge base: 1) the quantitative prediction of the benefit of zinc on PWSCC at candidate plants and 2) the application of zinc at low-concentrations (<10 ppb) for the purpose of PWSCC initiation mitigation.
There is therefore a need to provide an apparatus and method to protect a pressurized water reactor from stress corrosion cracking wherein the application of materials to limit the stress corrosion cracking must be at low zinc concentration levels of less than 10 ppb.
There is an additional need to provide an apparatus and method to protect currently operating pressurized water reactors from stress corrosion cracking without unduly increasing radiation levels for workers and equipment associated with reactor water coolant systems.
There is also a need to provide an apparatus and method to protect currently operating pressurized water reactors from stress corrosion cracking in an economical and non-damaging way for the nuclear fuel present in the reactor.